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Journal Articles

Development and validation of evaluation method on hypothetical total instantaneous flow blockage in sodium-cooled fast reactors and its application to a middle size SFR

Fukano, Yoshitaka

Proceedings of 25th International Conference on Nuclear Engineering (ICONE-25) (CD-ROM), 10 Pages, 2017/07

An evaluation on the consequences of a hypothetical total instantaneous flow blockage at the coolant inlet of an SA (HTIB) using SAS4A code was also performed in the past study. SAS4A code was further developed by implementing analytical model of power control system in this study. An evaluation on the consequences of HTIB in Monju by this developed SAS4A code was performed. It was clarified by the analyses considering power control system that the reactor would be safely shut down by the plant protection system triggered by either of 116 percent over power or delayed neutron detector trip signals. Therefore the conclusion in the past study that the consequences of HTIB event would be much less severe than that of unprotected loss-of-flow event was strongly supported by this study. Furthermore SAS4A code was newly validated using an in-pile experiment which simulated HTIB events. The validity of SAS4A application to safety evaluation on the consequence of HTIB was further enhanced in this study.

Journal Articles

A Study on the Promotion of Nuclear Security Culture

Tamai, Hiroshi; Tazaki, Makiko; Kokaji, Lisa; Shimizu, Ryo; Suda, Kazunori

Kaku Busshitsu Kanri Gakkai (INMM) Nippon Shibu Dai-36-Kai Nenji Taikai Rombunshu (Internet), 7 Pages, 2015/12

In recent years the promotion of nuclear security culture aiming at strengthening nuclear security is extensively mentioned, however, awareness of nuclear security culture seems to be not much high compared to the permeation of nuclear safety culture. As a world's leading country of peaceful nuclear use, permeation of nuclear security culture into each personnel attitude must be one of important issues in Japan. Learning from the TEPCO Fukushima Daiichi Nuclear Power Plant accident, complementarity between nuclear safety and nuclear security in the aspect of both protection measures has been profoundly recognised. Therefore, it will be natural to promote nuclear security culture modelled on the preceding nuclear safety culture. On this standpoint, the paper examines an approach for the promotion of nuclear security culture which, for example, consists of awareness cultivation, attitude progress, permeation assessment, and resulting in the establishment of PDCA Cycle.

JAEA Reports

Assessment report of research and development on "Nuclear Safety Research" in FY2014 (Post- and pre-review report)

Kudo, Tamotsu; Onizawa, Kunio*; Nakamura, Takehiko

JAEA-Evaluation 2015-011, 209 Pages, 2015/11

JAEA-Evaluation-2015-011.pdf:10.36MB

Japan Atomic Energy Agency (JAEA) consulted an assessment committee, "Evaluation Committee of Research and Development (R&D) Activities for Nuclear Safety", for post- and pre-review assessment of R&D on nuclear safety research. In response to JAEA's request, the Committee assessed mainly the progress of the R&D project according to guidelines, which addressed the rationale behind the R&D project, the relevance of the project outcome and the efficiency of the project implementation during the period of the current and next plan. As a result, the Committee concluded that the progress of the R&D project is satisfactory. This report describes the results of evaluation by the Committee. In addition, the appendix of this report contains presentations used for the evaluation, and responses from JAEA on the comments from the member of the Committee.

Journal Articles

Recent progress in research and development at Japan Atomic Energy Research Institute

Noda, Kenji; Yokota, Wataru

Denki Hyoron, 88(2), p.55 - 63, 2003/02

no abstracts in English

Journal Articles

Training of instructors on nuclear safety in Asian countries

Ikuta, Yuko; Shitomi, Hajimu*; Saeki, Masakatsu

Radioisotopes, 51(11), p.509 - 521, 2002/11

no abstracts in English

Journal Articles

Transfer of nuclear safety culture to Asian countries by international cooperation

Ikuta, Yuko; Shitomi, Hajimu*; Saeki, Masakatsu

Nippon Genshiryoku Gakkai-Shi, 44(10), p.744 - 745, 2002/10

no abstracts in English

Journal Articles

"Environmental Safety", ""Environmental Safety and Research Plan" and "Environmental Monitoring"

Kobayashi, Hideo

Genshiryoku Nenkan 2001/2002-Nen Ban, p.103 - 105, 2001/11

no abstracts in English

JAEA Reports

None

JNC-TN1400 2001-015, 509 Pages, 2001/10

JNC-TN1400-2001-015.pdf:25.67MB

no abstracts in English

JAEA Reports

None

JNC-TN1400 2001-014, 437 Pages, 2001/10

JNC-TN1400-2001-014.pdf:23.1MB

no abstracts in English

JAEA Reports

The calculation and estimation of wastes generated by decommissioning of nuclear facilities

; ; ; Takeda, Seiichiro

JNC-TN8420 2001-008, 134 Pages, 2001/07

JNC-TN8420-2001-008.pdf:4.4MB

This investigation was conducted as a part of planning the low-level radioactive waste management program (LLW management program). The aim of this investigation was contributed to compile the radioactive waste database of JNC's LLW management program. All nuclear facilities of the Tokai works and Ningyo-toge Environmental Engineering Center were investigated in this work. The wastes generated by the decomissioning of each nuclear facility were classified into radioactive waste and others (exempt waste and non-radioactive waste), and the amount of the wastes was estimated. The estimated amounts of radioactive wastes generated by decomissioning of the nuclear facilities are as follows. (1)Tokai works. The amount of waste generated by decommissioning of nuclear facilities of the Tokai works is about 1,079,100 ton. The amount of radioactive waste is about 15,400 ton. The amount of exempt waste and non-radioactive waste is about 1,063,700 ton. (2)Ningyo-toge Environmental Engineering Center. The amount of waste generated by decommissioning of nuclear facilities of Ningyo-toge Environmental Engineering Center is about 112,500 ton. The amount of radioactive waste is about 7,800 ton. The amount of exempt waste and non-radioactive waste is about 104,700 ton.

JAEA Reports

JAEA Reports

None

; *; *

JNC-TN8200 2001-001, 42 Pages, 2001/01

JNC-TN8200-2001-001.pdf:3.16MB

None

JAEA Reports

None

JNC-TN1400 2000-012, 250 Pages, 2000/11

JNC-TN1400-2000-012.pdf:10.18MB

no abstracts in English

JAEA Reports

None

JNC-TN1400 2000-010, 70 Pages, 2000/10

JNC-TN1400-2000-010.pdf:2.87MB

no abstracts in English

Journal Articles

''Environmental Safety'', ''Environmental Safety and Research Plan'' and ''Environmental Monitoring''

Nishiza, Masahiro

Genshiryoku Nenkan 2000/2001-Nen Ban, p.104 - 106, 2000/10

no abstracts in English

JAEA Reports

None

JNC-TN4420 2000-009, 11 Pages, 2000/06

JNC-TN4420-2000-009.pdf:0.84MB

None

JAEA Reports

ICONE-8 participation and investigation report of dry process in Argonne National Laboratory (ANL), USA

; Washiya, Tadahiro;

JNC-TN8420 2001-009, 48 Pages, 2000/04

JNC-TN8420-2001-009.pdf:0.58MB

ICONE(International Conference on Nuclear Engineering) is an international conference on nuclear chemical engineering held among the United States, Japan and Europe, and ICONE8 (the 8th time of the conference) was held at Baltimore, USA on April 2 to 6, 2000. The authors of this paper reported the latest information on the reprocessing technology in the following session of the conference and audited the panel discussion and the technical report of the dry reprocessing technology etc. in the conference. (1)Investigation of Safety Evaluation Method and Application to Tokai Reprocessing Plant (TRP) in session of Track-5 "Non-reactor Safety and Reliability" (Nakamura) (2)Structural Improvement on the continuous rotary dissolver in session of Track-9 "Spent Nuclear Fuel and Waste Processing" (Washiya) (3)Development of Evaporators Made of Ti-5% Ta Alloy and Zr - Endurance Test By Mock-Up unit" in session of Track-2 "Aging and Modeling of Component Aging, Including corrosion of Metals and Welds.. passivation, and passive films" (Takata) At the conference, about 650 people participated from the United States, Japan, France, Canada and others, about700 research announcements, 7 keynote lecture and 8 panel discussion were done, flourishing with many participants. Moreover, as the conference was held in the year of 2000, the evaluation of this century and the direction of the next century of nuclear energy were discussed. After the conference, authors visited Argonne National Laboratory (ANL-E, ANL-W) and exchanged information concerning dry process with researchers of ANL-E and ANL-W, visiting ANL facilities. It was very significant to be able to acquire the information on the dry process developed in ANL and realize the device scale and the development environment, etc. and acquire technical information in detail which would not be able to obtain by engineering data, exchanging information with ANL engineers directly. It is suggested to be very valuable that the ...

JAEA Reports

Examination of safety design guideline; Safety objective and elimination of re-criticality issues

; ; *;

JNC-TN9400 2000-043, 23 Pages, 2000/03

JNC-TN9400-2000-043.pdf:1.1MB

ln the feasibility study on commercialized fast breeder reactor (FBR) cycle systems conducted in JNC, it is required for candidate FBR plants that the level of safety should be enhanced so as to assure: (1)Comparative or superior safety level to that of light water reactors (LWRs), and (2)releaf of the public from anxiety about potential nuclear hazard. Adopting Passive safety characteristics is one of the measures. To attain the above safety objective, we considered implication of the basic safety principles for nuclear power plants that were created by the international nuclear safety advisory group of IAEA. The way to relieve from the anxiety was also taken into account. Then a definite safety objective was set from the standpoint of prevention of core disruptive accident (CDA). Furthermore, as a definite safety goal relating to reactor coresafety, elimination of re-criticality issues under CDA was set by considering characteristics of FBR in comparison with those of LWR. To examine measures for elimination of re-criticality issues, we developed a quick method to estimate possibility of re-criticality under CDA, by drawing a map about criticality characteristics under CDA in various degraded cores. Then hopeful measures were proposed for elimination of re-criticality issues in sodium-cooled FBR with mixed-oxide fuel. Molten fuel discharge behavior of their measures was preliminarily analyzed. We concluded that discharge capability of "a subassembly with an internal duct" was effective, and that "partial removal of axial blanket" was also effective as one of the measures though it has small effect on core performance.

JAEA Reports

None

JNC-TJ4420 2000-002, 794 Pages, 2000/03

JNC-TJ4420-2000-002.pdf:23.42MB

no abstracts in English

JAEA Reports

None

JNC-TJ4420 2000-001, 504 Pages, 2000/03

JNC-TJ4420-2000-001.pdf:17.19MB

no abstracts in English

116 (Records 1-20 displayed on this page)